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DOE OSTI2021

Modeling Activities Related to Waste Form Degradation: Progress Report

This report represents the milestone deliverable M4SF-21SN010309021 “Modeling Activities Related to Waste Form Degradation: Progress Report” that describes the progress of R&D activities of ongoing modeling investigations specifically on nuclear waste glass degradation, Density Functional Theory (DFT) studies on clarkeite structure and stability, and electrochemical modeling of spent nuclear fuel (SNF). These activities are part of the newly-created Waste form Testing, Modeling, and Performance work package at Sandia National Laboratories (SNL). This work package is part of the “Inventory and Waste Form Characteristics and Performance” control account that includes various experimental and modeling activities on nuclear waste degradation conducted at Oak Ridge National Laboratory (ORNL), SNL, Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL).

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DOE OSTI2021

GDSA Framework Development and Process Model Integration FY2021

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DOE OSTI2021

2021 Microbial U (Summary Report)

The US nuclear industry is on the cusp of significant growth in the areas of advanced reactors, SMRs, and microreactors. The existing commercial nuclear reactor fleet is considering expanding the life of the reactor cores by slightly increasing enrichment of their fuel. The common thread is the demand for High Assay Low Enriched Uranium (HALEU). A technical challenge exists with a blend down of Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) to meet the American Society for Testing and Materials (ASTM) and the “Y12” specifications. U-236 impacts reactor performance and is not desired in fresh fuel cores. In order to meet tight specifications, U-236 must be diluted by addition of fresh HEU. Fresh HEU is highly sought after and not plentiful in the US. A solution to this issue lies in the application of a technology to selectively remove the U-236 with non-hazardous microorganisms, which will be evaluated in this project by experimentally separating different isotopes of uranium.

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DOE OSTI2021

Uncertainty and Sensitivity Analysis Methods and Applications in the GDSA Framework (FY2021)

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (FCT) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). Two high priorities for SFWST disposal R&D are design concept development and disposal system modeling. These priorities are directly addressed in the SFWST Geologic Disposal Safety Assessment (GDSA) control account, which is charged with developing a geologic repository system modeling and analysis capability, and the associated software, GDSA Framework, for evaluating disposal system performance for nuclear waste in geologic media. GDSA Framework is supported by SFWST Campaign and its predecessor the Used Fuel Disposition (UFD) campaign. This report fulfills the GDSA Uncertainty and Sensitivity Analysis Methods work package (SF-21SN01030404) level 3 milestone, Uncertainty and Sensitivity Analysis Methods and Applications in GDSA Framework (FY2021) (M3SF-21SN010304042). It presents high level objectives and strategy for development of uncertainty and sensitivity analysis tools, demonstrates uncertainty quantification (UQ) and sensitivity analysis (SA) tools in GDSA Framework in FY21, and describes additional UQ/SA tools whose future implementation would enhance the UQ/SA capability of GDSA Framework. This work was closely coordinated with the other Sandia National Laboratory GDSA work packages: the GDSA Framework Development work package (SF-21SN01030405), the GDSA Repository Systems Analysis work package (SF-21SN01030406), and the GDSA PFLOTRAN Development work package (SF-21SN01030407). This report builds on developments reported in previous GDSA Framework milestones, particularly M3SF 20SN010304032.

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DOE OSTI2021

Dual Purpose Canister Reactivity and Groundwater Absorption Analyses (Rev. 7)

The current spent nuclear fuel (SNF) management strategy includes reliance on dry storage. Utilities are meeting their interim storage needs on an individual basis with use of large-capacity dry storage casks, with a current focus on meeting existing storage and transportation requirements, as disposal requirements are not currently available. These casks are commonly known as dual-purpose (i.e., storage and transportation) canisters (DPCs). However, a small percentage of single-purpose (storage only) systems is also being used to meet storage needs. These are included under the “DPC” heading. This report investigates the postclosure criticality safety aspects of DPCs.

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DOE OSTI2022

PFLOTRAN Development FY2022

The Spent Fuel & Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is to develop a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media. This report describes fiscal year (FY) 2022 accomplishments by the PFLOTRAN Development group of the SFWST Campaign. The mission of this group is to develop a geologic disposal system modeling capability for nuclear waste that can be used to probabilistically assess the performance of generic disposal concepts. In FY 2022, the PFLOTRAN development team made several advancements to our software infrastructure, code performance, and process modeling capabilities.

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DOE OSTI2022

Uncertainty and Sensitivity Analysis Methods and Applications in the GDSA Framework (FY2022)

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DOE OSTI2022

Review of Residual Stresses in Spent Nuclear Fuel Canisters

This report compiles the set of published results on the welding residual stress (WRS) distributions across the thickness section at the welds in spent nuclear fuel (SNF) canisters before and after a repair is done. The WRS information for similar steel structures with thick (~ 12.7 mm (0.5 inch)) large plate construction are also compiled to provide a comprehensive set of information on: methods to quantify; modeling to estimate; and methods to reduce WRS.

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DOE OSTI2022

Ageing Management for Extended Long-Term Dry Storage of Spent Nuclear Fuel and Transportation

The principal objectives of the Coordinated Research Project (CRP) and CRADA were to (1) investigate how ageing effects leading to degradation of materials used in the spent fuel dry storage systems could be managed by ageing management programs (AMPs) and (2) using existing AMPs as a basis, establish guidance on how to develop, generate, and maintain AMPs for dry storage systems of spent nuclear fuel (SNF) that can be accomplished in various ways. The PI (Dr. Liu) is an internationally recognized expert on ageing management for license renewal of nuclear power plants and independent spent fuel dry storage installations. Dr. Liu was invited by IAEA to participate in this CRP and served as Chair of the Working Group, supporting the IAEA CRP Lead, with other members from Argentina, the Czech Republic, France, Germany, Hungary, Japan, Pakistan, Spain, Switzerland, United Kingdom, and the United States of America. Argonne’s scope of work in the CRADA for the CRP included (1) ageing management guidance documents developed by Argonne for DOE and used by the Nuclear Regulatory Commission (NRC) and industry and (2) research leveraged from the DOE Office of Nuclear Energy on the mechanical properties of high-burnup fuel cladding and the ARG-US remote monitoring systems technology developed for the DOE Packaging Certification Program, Office of Packaging and Transportation, Office of Environmental Management.

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DOE OSTI2022

GDSA Framework Development and Process Model Integration FY2022

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is disposal system modeling (Sassani et al. 2021). The SFWST Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media. This report describes fiscal year (FY) 2022 advances of the Geologic Disposal Safety Assessment (GDSA) performance assessment (PA) development groups of the SFWST Campaign. The common mission of these groups is to develop a geologic disposal system modeling capability for nuclear waste that can be used to assess probabilistically the performance of generic disposal options and generic sites. The modeling capability under development is called GDSA Framework (pa.sandia.gov). GDSA Framework is a coordinated set of codes and databases designed for probabilistically simulating the release and transport of disposed radionuclides from a repository to the biosphere for post-closure performance assessment. Primary components of GDSA Framework include PFLOTRAN to simulate the major features, events, and processes (FEPs) over time, Dakota to propagate uncertainty and analyze sensitivities, meshing codes to define the domain, and various other software for rendering properties, processing data, and visualizing results.

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DOE OSTI2022

Deposition Modeling Updates for a Vertical System

This report provides deposition modeling updates and results for a vertical spent nuclear fuel (SNF) storage system. The vertical storage system was the NAC International’s Modular, Advanced Generation, Nuclear All-purpose STORage System (MAGNASTOR®). The updates to the model included running a turbulent model sensitivity study and incorporating droplet evaporation into the deposition model for the MAGNASTOR®. In addition to the updates to the vertical MAGNASTOR system this report discusses the topic of particle resuspension and future work related to looking at resuspension in both a vertical and horizontal storage system.

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DOE OSTI2022

INL Site ARG-US Implementation – FY22 Activities and FY23 Plans

The Packaging Certification and Life Cycle Management program at Argonne National Laboratory (ANL) developed a suite of monitoring systems collectively referred to here as ARG-US. ARG-US provides necessary data for nuclear facility and system operation and maintenance, and has been previously demonstrated in hot cells, radioactive material (RAM) storage areas, and RAM shipment trucks. It has also been installed directly on RAM storage and shipment containers. ARG-US offers some unique advantages over other commercially available systems by using wireless data connections, battery power supplies, and customizable monitoring methods. Idaho National Laboratory (INL) has been tasked with investigating applications for ARG-US at INL site facilities, which are operated by several different contractors. The initial investigation scope centered on CPP-603 Irradiated Fuel Storage Facility in relation to the upcoming Department of Energy (DOE) Spent Nuclear Fuel (SNF) Packaging Demonstration. The investigation has been led by the Used Fuel Management Department in INL’s Nuclear Science and Technology (NS&T) directorate. This investigation is expected to recommend INL site processes or facilities in which to implement ARG-US systems. INL has engaged local site technical and oversight representatives, security, nuclear safeguards, and program management personnel to identify good candidates for ARG-US test implementations. INL has identified the following high-level goals for any INL site implementation of ARG-US: provide unique testing environments, prompt development of new monitoring methods/techniques for the ARG-US suite, and acquire useful monitoring for the user facility. As a result, INL suggests three program areas for further investigation: CPP-603 Fuel Handling Cave (FHC), legacy mixed waste storage systems at Idaho Nuclear Technology & Engineering Center (INTEC) known as the “Tank Farm,” and periodic and emergency environmental monitoring. This progress report relates activities undertaken in this investigation, describes the preliminary areas of interest for limited scope ARG-US testing or implementation, and relays expected actions for completing the task scope. This report offers an opportunity to the program sponsor, technical leads at ANL, and INL site representatives to give feedback on the initial assessment and make recommendations on the forthcoming activities.

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DOE OSTI2022

Corrosion of U233-Doped Uranium Oxide using Microfluidics Methods

The aim of work this year has been to investigate the role of alpha (α)-radiation and hydrogen (H 2 ) on the corrosion of uranium oxide (UO 2 ) using a microfluidic device. The microfluidic device, termed the Particle-Attached Microfluidic Electrochemical Cell, (PAMEC), enables monitoring of the UO 2 electrochemical corrosion potential (E corr ) under different environments, including de-aerated conditions and in the presence of dissolved H 2 . The Si 3 N 4 window allows us to study morphological and chemical changes under an electron microscope. We have previously demonstrated that the PAMEC matches the results from bulk electrochemical tests with UO 2 [1]. We used the high specific activity uranium (U) isotope, 233 U, (t 1/2 = 160,000 years) incorporated into UO 2 , to generate a localized a-field. The objective of the experiments were to mimic the radiation environment that would be experience at the surface of aged spent nuclear fuel (SNF) during long-term geologic disposal under anoxic conditions. Wittman et al. [2] predicted that in the presence of a pure a-radiation field and under H 2 conditions, the concentration of the radiolytic oxidant H 2 O 2 would be suppressed or even eliminated. In a UO 2 corrosion experiment this would be exhibited through a lowering of the measured UO 2 corrosion potential compared to identical conditions in the absence of dissolved H 2 . We found that the predictions of Wittman and co-workers were supported and that the corrosion potential of the 233 U-doped UO 2 in solution lowered with presence of H 2 gas and increased in the absence of H 2 , when under anoxic conditions. The PAMEC experiments indicate that H 2 O 2 has been eliminated in a solution sparged with H 2 while exposed to an a-radiation field. The corrosion potential of the a-doped 233 U(10%)- 238 UO 2 in a solution sparged with Ar/H 2 matched the corrosion potential of 238 UO 2 in solution sparged with air. This clearly demonstrated that the H 2 O 2 had been eliminated and that the only oxidant present was O 2 in this system in complete agreement with the modeling results of Wittman et al. These results point to the need to improve the Fuel Matrix Degradation (FMD) Process Model training data set that is being used in the FMD surrogate model that is being developed for the repository program. The incorporation of realistic radiation chemistry will improve the scientific basis for the FMD Model.

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DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water

The leaching experiment aims to understand the trends in the radiolysis-enhanced dissolution of HBU SNF when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL deionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed using gamma spectroscopy and inductively coupled plasma - mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of the fuel exposed.

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DOE OSTI2023

A Vacuum Drying Study of Simulated Failed Nuclear Fuel (FY23)

This report documents work performed under the Office of Spent Fuel and Waste Disposition’s (SFWD’s) Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 4 milestone M4SF-23OR010203013. Over the 70 years of light-water reactor (LWR) operation, fuel rod failures have occurred during reactor operation. Based on the Unified Database that SFWD maintains, there are ~5,400 LWR fuel rods that developed leaks during reactor operation of the >19 million rods operated. The Unified Database indicates that ~4,200 of these rods remain in the spent fuel pool, and ~1,200 are in dry storage canisters. It is expected that all failed rods are “waterlogged” in the spent fuel pool, meaning the inside of the rods contain some volume of water. There are two related objectives of this work. The first is to develop a simple and inexpensive experiment to better understand removal of water from inside failed spent nuclear fuel (SNF) rods, and the second is to develop a model to predict drying of simulated failed fuel in the simple experiment. Once a model captures the phenomenon in the simple experiment, it can be extended to predict drying under more prototypical conditions.

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DOE OSTI2023

Evaluation of Flaw Detection with SLDV Methods Before/After Mechanical Patch

This report provides a summary of the testing and the experiments to evaluate the utility of laser ultrasonic testing (UT) in detecting and characterizing defects on the surface of steel plates before and after the application of an engineered composite patch for crack repair. Included are the detection and characterization of defects from a section of welded plate material taken from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister which had been pre-flawed and exposed to deliquescent sea salt conditions for 2 years. The information includes a review of previously reported metallurgical characterization, and new work on detailed surface characterization of the corrosion attack using laser UT.

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DOE OSTI2023

Management of Alkali and Alkaline Earth Fission Products in Used Pyroprocessing Salt

Pyroprocessing of spent nuclear fuel (SNF) involves dissolving metallic fuel into a molten salt electrolyte (typically eutectic LiCl-KCl) and then preferentially depositing actinides onto inert cathodes. Subsequent operations include drawdown of residual actinides and lanthanides from the electrolyte prior to re-using the salt. The recovered actinides are recycled and the recovered lanthanides are disposed as waste. Alkali and alkaline earth metal fission products in the fuel, such as Cs, Sr and Ba, dissolve into the salt during electrorefining. The concentrations of these elements buildup over time in the molten salt electrolyte, which may change the freezing point. The radioactive decay of 137 Cs and 90 Sr (half life 30 and 29 years) generates significant heat and produces strong ionizing radiation fields (β and γ). The increasing heat load and radioactivity as these elements build up in the molten salt requires frequent replacement and disposal of the electrolyte salt. Alternatively, the salt can be treated to remove these and other elements and then recycled to the electrorefiner. An effective strategy to manage these alkali and alkaline earth metal fission products in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. Alkali and alkaline earth metal fission products are extremely stable in molten salt as chlorides--even more stable than the LiCl-KCl eutectic base salt--making them challenging to remove. They are not removed during drawdown operations to recover residual actinides and lanthanides and a separate operation is required to sufficiently purify the salt for reuse. This work is focused on selecting a method for separating Cs, Sr and Ba from the salt recovered from the lanthanide drawdown operation prior to recycling the cleaned salt back to the electrorefiner. Not addressed in this work is the management of the waste stream produced by the separation. This report summarizes the issues to be addressed when developing removal strategies for cesium, strontium, and barium and reviews existing methods to identify suitable methods and any technological gaps in their application.

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DOE OSTI2023

Dual-Purpose Canister Filling Demonstration Project Progress Report at ORNL, 2023

The US DOE Office of Nuclear Energy is investigating the feasibility of direct disposal of dual-purpose canisters (DPCs) in a hypothetical geological repository to offset the potential requirement to repackage spent nuclear fuel (SNF) from existing DPCs into smaller, disposal-ready canisters. Oak Ridge National Laboratory (ORNL) is currently evaluating the feasibility of filling void space in loaded DPCs with an engineered material to prevent a criticality event caused by groundwater/moderator intrusion. Metal alloys are being investigated as a filler material because of their relatively low viscosities when molten, which may facilitate their injection via an existing drainpipe that runs almost the full length of the DPC. ORNL’s strategy for evaluating filler viability includes simulations and physical demonstrations of filling and casting behavior, as well as evaluations of materials for compatibility in the repository environment. Filling of DPCs in this manner is expected to mitigate the risk associated with a post-closure criticality event during the repository performance assessment time frame (10,000 years or greater). Efforts in this fiscal year focused on (1) destructive analysis of experimental filler castings made in FY 2022, (2) a report outlining a conceptual design of a DPC filling facility (Fortner et al., 2023, M3SF 23OR010305044/ ORNL/SPR-2023/2921, May 31, 2023), (3) a report on affected features, events, and processes (FEPs) due to DPC filling (Price et al., 2023, M3SF-23SN010305093, issuance pending), (4) developing and testing a more practical alloy filler based upon a Sn-Al eutectic, and (5) modeling the heating and cooling dynamics of a DPC subjected to molten metal filling. The preliminary results from each of these tasks support the feasibility of filling DPCs with metal as a strategy against the possibility of criticality in the repository. The FY 2022 casting was found to penetrate even very small orifices in the mold and internal structures. Preliminary testing of the Sn-Al eutectic indicate little interaction of the melt with Zircaloy cladding. Thermal modelling shows that a filled DPC will cool to manageable temperatures within 2-3 days, which is likely manageable in a facility.

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